ML20134L979

Proposed Tech Specs Re Relocation of Turbine Overspeed Protection

ML20134L979
Person / Time
Site:	Point Beach
Issue date:	02/12/1997
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20134L966	List: ML20134L962 ML20134L979
References
NUDOCS 9702200042
Download: ML20134L979 (7)

Text

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Attachment 4 Marked-Un Technical Specifications Panes

Technical Specifications Channe Reauests 1%

Relocation of Turbine Overspeed Prvtection Technical Specifications to FSAR Point Beach Nuclear Plant. Units 1 and 2 i

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9702200042 970212 PDR ADOCK 05000266-P PDR

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2. Single Unit Operation - One of the three operable auxiliary feedwater pumps associated with a unit may be out-of-service for the below specified times. The turbine driven auxiliary feedwater pump may be out-of-service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the turbine driven auxiliary feedwater pump cannot be restored to service within that j

-72 hour time period, the reactor shall be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1 Either one of the two motor driven auxiliary feedwater pumps rnay be out-of-service for up to 7 days._ If the motor driven auxiliary feedwater pump cannot be restored to service within that 7 day period the operating unit shall be in hot shutdown within the  !

l next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D. The main steam stop valves (MS-2017 and MS-2018) and the non-return check valves (MS-  !

l l 2017A and MS-2018A) shall be operable. If one main steam stop valve or non-retum check i l valve is inoperable but open, power operation may continue provided the inoperable valve is  ;

1 restored to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise the reactor shall be placed in a hot shutdown condition within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With one or more main steam stop valves i

or non-return check valves inoperable, subsequent operation in the hot shutdown condition may proceed provided the inoperable valve or valves are maintained closed. An inoperable main steam stop valve or non-return check valve may however, be opened in the hot L

shutdown condition to cool down the affected unit and to perform testing to confirm  !

operability.

E The crc crer s: cam dump system cha!! be operable. !f the crossover steam dump s,3:e:wis determined ic be inoperable, reduce pe".er-to less than 180 MWe (grc 3) "ithin 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />s-F During pe ver operati(m, at least one ef the turbine overspeed protectien systems that trip 4he turbine ::cp valves er -hut the turke ge.enw+alves cha!!4wrable. If all three :; : ems are determmed c be inoperah!e,ine! ate the turbine from the 3:eam supply cithkwhe-neu-six

! hours-

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l Unit 1 Amendment NoA41 A4areh44W5 Unit 2 - Amendment NoA65 15.3.4-2a l

i i d G. Should one ef the turMne step va!ves er govemer .alves he declared inoperable, restore the inoperable valve to an operable s:atus "ithin-72 h- eure-4foperaNiity-eannot be restarah perform +neef-the fc!!cv ing actions

! Shut the affected va!ve within the next ,ix-heure l

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2. belate the turbmefrenHhe steam supp!; "it!Hhe next six hours.

Basis A reactor shutdown from power requires removal of core decay heat. Immediate decay heat removal requirements are normally satisfied by the steam bypass to the condenser. Therefore, core decay heat can be continuously dissipated via the steam bypass to the condenser as feedwater in the steam generator is converted to steam by heat absorption. Normally, the capability to return feedwater flow to the steam generators is provided by operation of the turbine cycle feedwater system.

The eight main steam safety valves have a total combined rated capability of 6,664,000 lbs/hr. The total full power steam flow is 6,620,000 lbs/hr, therefore eight (8) main steam safety valves will be able to relieve the total full-power steam flow if necessary.

l 1

1 In the unlikely event of complete loss of electrical power to the station, decay heat removal would continue to be assured for each unit by the availability of either the steam-driven auxiliary feedwater pump or one of the two motor-driven auxiliary steam generator feedwater pumps, and steam discharge to the atmosphere via the main steam safety valves or atmospheric relief valves. One motor-driven auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from a unit. The minimum amount of water in the condensate storage tanks ensures the ability to maintain each unit in a hot shutdown condition for at least one hour concurrent with a loss of all AC power.

An unlimited supply is available from the lake via either leg of the plant sersice water system for an indefinite time period.

Unit 1 - Amendment No.-447 (4*reetion4etk+ dated Unit 2 - Amendment No.-14 15.3.4-2b May !!,1991

Each of the AFW pumps possesses a low suction pressure trip that will protect it should a loss of l feedwater occur. Additionally, should a steam generator tube rupture occur, the motor-operated steam admission valves for the turbine-driven AFW pumps serve as isolation boundaries for the affected steam generator.

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The atmospheric steam dump lines are required to be operable because they are relied upon, following a steam generator tube rupture coincident with a loss of A.C. power, to cool down the Reactor Coolant System to RHR entry conditions. An atmospheric steam dump line is considered operable ifit is capable of providing the controlled relief of main steam flow necessary to perform the RCS cooldown. Isolating an atmospheric steam dump line does not render it inoperable if the line can be unisolated and the RCS can still be cooled down to RHR entry conditions, through local or remote operation, within the time period required by the applicable FSAR accident analyses.

Unit 1 - Amendment No.447 Correetion4etter4ated Unit 2 - Amendment No.4M 15.3.4-2c Ma.t !!.1991

TABLE 15.4.1-1 (continued) ~

PLANT CONDITIONS NQ, CIIANNEL DESCRIPTION CIIECK CALIBRATE TEST WilEN REOUIRED

36. Radiation Monitoring System D(7) R(7) M(7) ALL

37. Reactor Vessel Fluid Level System M R -

ALL

38. Refueling Water Storage Tank Level -

R -

ALL

39. Residuallleat Removal Pump Flow -

R -

ALL

40. Safety Valve Position Indicator M R -

ALL

41. Subcooling Margin Monitor M R -

ALL 42 TwS:n Over: peed T:ips Deleted

'ndependen: Oven: peed Prc:ee&n Sy; ten 't M(!) ^lA Os en peed !!!cck trip R M(!) A!1

43. Volume Control Tank Level -

R -

ALL

44. Reactor Protection System and - - M(1,23) ALL Emergency Safety Feature Actuation System Logic

45. Reactor Trip System Interlocks

-Intermediate Range Neutron Flux, P-6 -

R(24) R ALL

-Power Range Neutron Flux, P-8 -

R(24) R ALL

-Power Range Neutron Flux, P-9 -

R(24) R ALL

-Power Range Neutron Flux, P-10 -

R(24) R ALL

-Ist Stage Turbine impulse Pressure -

R(24) R ALL Unit 1 - Amendment No.45-7 Page 4 of 6 Deeember 8,1991 Unit 2 - Amendment No-M4

, = ~ . . - - - . ..

l TABLE 15.4.1-2 (Continued)

'. 1 Icst Freauenev

7. Spent Fuel Pit a) Boron Concentration Monthly b) Water Level Verification Weekly

8. Secondary c' a' ant Gross Beta-gamma Weekly (')

Activity or gamma  !

isotopic analysis I I

lodine concentration Weekly when gross l

Beta-gamma activity l

equals or exceeds 1.2 Ci/cc

9. Control Rods a) Kod drop times of all Each refueling or fulllength rodsU) after maintenance that could affect i proper functioning ") I b) Rodworth measurement Following each refueling shutdown prior to commencing power operation

10. Control Rod Partial movement of Every 2 weeks 04 all rods i1. Pressurizer Safety Valves Set point Every five years '")

12. Main Steam Safety Valves Set Point Every five years (")

13. Containment isolation Trip Functioning Each refueling shutdown

14. Refueling System Interlocks Functioning Each refueling shutdown

15. Service Water System Functioning Each refueling shutdown J

16. Primary System Leakage Evaluate Monthly )

17. Diesel Fuel Supply Fuel inventory Daily

18. Twbirdopand4ewrnor_Ddeled Func4kming A:maalV' Valws

19. L: ; Pr=ure Tu:hineJdeled V !' s." and4nagnene Rwry4ws ReiOF4n$peCliOR- par!!chW4N}HN}

penetrant

20. Boric Acid System Storage Tank and Daily"*

piping temperatures  !

2 temperature required l by Table 15.3.2-1 I

Unit 1 - A.mendment No.448 Unit 2 - Amendment No.-442 Page 2 of 4 December 12,19%

I

.- - .-. . . _ - = . - - -. . - - . .-. . . . - .

, TABl.E 15.4.1-2 (Continued)

Icil Freauency

21. PORV Block Valves a. Complete Valve Cycle Quarterly (")

b. Open position check Every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ("' >

22. Integrity of Post Accident Evaluate Each refueling ,

Recovery Systems Outside cycle Containment l

l 23. Containment Purge Supply Verify valves are Monthly M l l and Exhaust isolation locked closed l Valves l

24. Reactor Trip Breakers a. Verify independent Monthly

• J operability of automatic shunt and undervoltage trip functions. i i

b. Verify independent Each refueling operability of man- shutdown ual trip to shunt and undervoltage trip functions.

25. Reactor Trip Bypass a. Verify operability Prior to Ikeakers of the undervoltage breaker use trip function.

b. Verify operability Each refueling of the shunt trip shutdown functions.

c. Verify operability Each refueling of the manual trip shutdown to undervoltage trip functions.

26. 120 VAC Vital Instr. Verify Energizedo2) Shiftly ,

Bus Power i l

27. Power Operated Relief Operate"0 Each shutdown""

Valves (PORVs),

PORV Solenoid Air Control Valves, and Air System Check

28. Atmospheric Steam Dumps Complete valve cycle Quarterly

29. Gn= :: Stea : D=7 ys:::n_Dekted--Vwifyqwal444 S y e Quarterly eadH,teanulemp+alves Unit 1 - Amendment No.-M8 Unit 2 - Amendment No.442 Page 3 cf 4 Deeember 12,1991